Some experiences and development in Japan related to nondestructive examination and fracture mechanics in structural integrity assessment of nuclear power plant components

1991 ◽  
Vol 131 (3) ◽  
pp. 329-336 ◽  
Author(s):  
N. Maeda ◽  
G. Yagawa
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Fracture Mechanics ◽  
Nuclear Power ◽  
Structural Integrity ◽  
Nondestructive Examination ◽  
Integrity Assessment ◽  
Plant Components

What’s New in Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components

2004 ◽  
Author(s):  
Gary Park
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Mechanical Engineering ◽  
Nuclear Industry ◽  
Nondestructive Examination ◽  
American Society ◽  
Plant Components

The nuclear industry is a pretty dynamic industry, in that it is always on the move, changing every time we turn around. For that very reason, there is a need to keep up with the industry by providing changes to American Society of Mechanical Engineering Section XI, “Rules for Inservice Inspection of Nuclear Power Plant Components.” There have been many changes over the last three years. This paper addresses a few of those, but gives a feel for the number of changes from the 2000 Addenda to the 2003 Addenda, there have been a total of approximately 56 changes. Of those changes, 11 were in the repair/replacement requirements, 19 in the inspection requirements, 4 in the evaluation requirements, 18 in the nondestructive examination requirements, and 4 in the administrative requirements. The paper classifies the changes as “Technically Significant,” “Significant,” “Non-Significant,” or “Editorial.” The paper addresses only a few of those changes that were “Technically Significant.” The paper also includes some of the activities that the ASME Section XI Subcommittee is currently working on.

Study on Load Capacity of Main Steam Superpipe Nozzle Used in Nuclear Power Plant

2017 ◽  
Author(s):  
Zhou Gengyu ◽  
Liang Shuhua ◽  
Sun Lin ◽  
Lv Feng
Keyword(s):  
Finite Element ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Structural Integrity ◽  
Load Capacity ◽  
Operating Conditions ◽  
Integrity Assessment ◽  
Inner Radius ◽  
Main Steam

The main steam super pipe used in nuclear power plant is an important safety class2 component. There are several nozzles located on it and linked with main steam safety valves. In the past two decades, the hot extrusion forming technology has been widely used to manufacture the super pipe nozzles. Comparing with traditional insert weldolet, the wall thickness of the extruded nozzle is relative small, and the nozzle inner radius is hard to control precisely in the fabrication process. Due to high temperature working condition and complicated loading conditions, the load capacity of the super pipe extruded nozzle has become an issue of concern for manufacturers and users. This paper presents a structural integrity assessment of a super pipe extruded nozzle. The nozzle stresses due to internal pressure and external loads for different operating conditions are obtained by the three-dimensional finite element analysis. The extruded nozzle is evaluated against the RCCM code Subsection C3200 Service Levels O, B and D stress limits for design, upset and faulted conditions. A parametric sensitivity analysis of the extruded nozzle inner radius size is also carried out. In addition, in order to reduce the calculation effort, an efficient calculation method is developed by using the commercial finite element program ANSYS.

Overview of the JEA Code of Fracture Toughness Requirements for Nuclear Power Plant Components

2006 ◽  
Author(s):  
Minoru Tomimatsu ◽  
Seiji Asada ◽  
Takashi Hirano ◽  
Hideo Kobayashi
Keyword(s):  
Fracture Toughness ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Energy Use ◽  
Evaluation Method ◽  
Structural Integrity ◽  
Charpy Impact ◽  
Test Methods ◽  
Plant Components

The Japan Electric Association Code, JEAC 4206, “Method of Verification Tests of the Fracture Toughness for Nuclear Power Plant Components” was originally published in 1973 in order to prescribe test methods, fracture toughness requirements and acceptance standards for materials used in nuclear power plant components. The code was recently revised so as to incorporate a new method to evaluate the structural integrity of reactor vessels with upper shelf Charpy impact energy (USE) less than 68J, based on the results of the researches performed as national projects by the Japan Power Engineering and Inspection Corporation and so on. In this paper, some contents of the code, which are applicable for reactor vessels, such as J integral based integrity evaluation method for reactor vessels with low USE including predicting J resistance curves (J-R curves) by using USE and temperature, and methods to evaluate integrity against pressurized thermal shock events and to determine pressure temperature limits, are overviewed.

Expanded Influence of UT Performance Demonstration and Fitness-for-Service Analysis in the Nuclear Power Industry

2011 ◽  
Author(s):  
Owen F. Hedden ◽  
C. David Cowfer
Keyword(s):  
Structural Analysis ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Pressure Vessel ◽  
Nuclear Power ◽  
Probability Of Detection ◽  
Nuclear Power Industry ◽  
Nondestructive Examination ◽  
Performance Demonstration ◽  
Plant Components

This paper addresses recent development and application of ASME Boiler and Pressure Vessel Code and Code Case requirements for weld nondestructive examination (NDE) and for fitness-for-service (FFS) structural analysis of the flaws thus discovered. A brief description of the basis for this development is included. Reviews are presented of additions to the literature addressing applications to nuclear power plant components. Issues regarding application of Code Cases 2235 and N-529 are presented. Concerns regarding the importance of reliable probability of detection (POD) data for different weld categories are addressed. Concerns with appropriate acceptance standards for flaws in different weld categories and materials are also addressed.

Advances in Residual Stress Modeling and Measurement for the Structural Integrity Assessment of Welded Thermal Power Plant

2008 ◽  
Vol 41-42 ◽  
pp. 391-400 ◽  
Author(s):  
Lyndon Edwards ◽  
Mike C. Smith ◽  
Mark Turski ◽  
Michael E. Fitzpatrick ◽  
P. John Bouchard
Keyword(s):  
Residual Stress ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Residual Stresses ◽  
Nuclear Power ◽  
Structural Integrity ◽  
Pressure Vessels ◽  
Economic Life ◽  
Life Assessment ◽  
Integrity Assessment

The safe operation of both thermal and nuclear power plant is increasingly dependent upon structural integrity assessment of pressure vessels and piping. Furthermore, structural failures most commonly occur at welds so the accurate design and remnant life assessment of welded plant is critical. The residual stress distribution assumed in defect assessments often has a deciding influence on the analysis outcome, and in the absence of accurate and reliable knowledge of the weld residual stresses, the design codes and procedures use assumptions that yield very conservative assessments that can severely limit the economic life of some plant. However, recent advances in both the modeling and measurement of residual stresses in welded structures and components open up the possibility of characterising weld residual stresses in operating plant using state-of–the–art fully validated Finite Element simulations. This paper describes research undertaken to predict residual stresses in stainless steel welds in order to provide validated reliable, accurate Structural Integrity assessment of nuclear power plant components

A Technical Basis for Characterizing Flaws Detected by Preservice and Inservice Examinations of Nuclear Power Plant Components

1975 ◽  
Vol 97 (4) ◽  
pp. 322-326 ◽  
Author(s):  
R. R. Maccary
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Nuclear Reactor ◽  
Flaw Detection ◽  
Nondestructive Examination ◽  
Asme Code ◽  
Underlying Basis ◽  
Technical Basis ◽  
Plant Components

The nondestructive examination procedures specified by the rules of construction of the ASME Boiler and Pressure Vessel Code—Section III, “Nuclear Power Plant Components” require techniques whose flaw detection capabilities are well within the practical limits established for acceptable workmanship and quality of fabrication. The rules of the ASME Section XI, “Inservice Inspection of Nuclear Reactor Coolant Systems”, impose an additional series of examinations. Material or fabrication flaws detected during a preservice examination as well as flaws developed during service must be evaluated to establish the acceptability of the component for initial and continued service. These examination requirements have introduced the need to characterize the flaws detected by the examinations and to set “allowable flaw indication standards.” The principles of fracture mechanics provide an engineering tool which predicts the behavior of materials containing flaws under service loadings. These principles form the underlying basis upon which the allowable flaw indication standards of ASME Section XI were formulated. The development of new rules governing flaw indication characterization and allowable flaw indications standards, as specified in the ASME Code, Section XI, are reviewed.

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